Compositions and methods for monitoring actinides

ABSTRACT

Compositions and methods for monitoring the quantity of actinides present in a test sample are disclosed. Compositions and methods for monitoring the motion of special nuclear materials through space are also described. Compositions and methods for monitoring the quantity of a fissile special nuclear material present in a test sample are disclosed. Compositions and methods for monitoring actinides during reprocessing of spent nuclear fuel after 30-year cool down are disclosed. Compositions and methods for monitoring actinides during reprocessing of spent nuclear fuel after 180 day cool down are also disclosed.

GOVERNMENT RIGHTS

This invention was made with government support under DE-FG07-07ID14890 awarded by the Department of Energy (DOE). The United States government has certain rights in the invention.

BACKGROUND OF INVENTION

The U.S. Department of Energy has developed methods for reprocessing spent nuclear fuel in commercial reprocessing plants. These methods present challenges within the nuclear industry with regard to security. The majority of this development was accomplished under the Advanced Fuel Cycle Initiative, building on the legacy of process research and development over the past 50 years. The emergence of the new Global Nuclear Energy Partnership aims to continue and expand the development of Advanced Fuel Cycle Initiative processing methods. This initiative has elevated the U.S. and worldwide commitment to advance fuel processing. These advanced processing methods need to be scaled up and engineered for real-scale implementation.

The most prominent processing method under development is named UREX+ depicted schematically in FIG. 1. The name actually refers to a family of processing methods that begins with the Uranium Extraction (UREX) process and incorporates a variety of other methods to separate uranium, selected fission products, and the transuranic isotopes from dissolved Spent Nuclear Fuel. UREX+ is similar to the well-known PuREX process currently used worldwide (e.g., Sellafield, Great Britain; La Hague in France; and Rokasho in Japan). The similarities derive from multiple chemical separation processes that are used to remove the major sources of radioactivity with specific goals to recycle U and Pu into the fuel cycle.

Processing will be needed for over 1,000 tons of fuel per day in the future to accommodate the worldwide Spent Nuclear Fuel from about 1,000 operating reactors each with an inventory of about 100 T of UO₂ and, additionally for the reprocessing of the legacy inventory of Spent Nuclear Fuel. As Global Nuclear Energy Partnership and the U.S. Department of Energy moves toward implementation of UREX+ over the next 20 years, strategies, material controls, and accountability methods will be required. Monitoring actinides with higher molecular weight during aqueous separations is a critical research area of the U.S. Global Nuclear Energy Partnership and Advanced Fuel Cycle Initiative programs. A key aspect of monitoring for material accountability is a method for assessing in real-time composition of the Spent Nuclear Fuel in order to detect possible diversion of Transuranic elements such as Pu. Such timely detection is especially important for ²³⁹Pu. A single fuel assembly can contain close to 7 kg, the quantity which is sufficient to produce Nagasaki-type nuclear explosives. Likewise, in the Material Balance Areas of critical nuclear installations (e.g., weapons plants), the tracking and inventory control is crucial for safeguarding and securing relevant nuclear materials. By providing an on-line material accounting system during the recycling processes, uninvited diversion of the material streams may be curbed. Currently, however, maintenance and control of such Special Nuclear Materials (SNMs) are conducted via time-consuming off-site monitoring and assessments, which significantly reduce the speed and efficacy of reprocessing and functioning of critical nuclear facilities.

Currently, alpha emitter detection requires time-consuming off-site laboratory based methods, and most on-line neutron detection systems are readily saturated in the extreme beta-gamma fields associated with the copious quantities of fission products like ¹³⁷Cs. As noted in FIG. 1, the first step in reprocessing at any reprocessing facility is to chop the Spent Nuclear Fuel (SNF) and dissolve the materials, known as accountable materials, in an acidic solution, which is accountable for the entirety of nuclear materials in the SNF. The accountable materials containing solution is then transferred to an accountability tank. Samples are taken and measurements are made to crudely determine the total quantity of initial nuclear material inventory. Determining the initial Special Nuclear Materials quantity is crucial in maintaining material accountability throughout the process and in identifying whether diversion, if any, has occurred.

Unfortunately, Near Real Time Accountability of transuranic actinides has not been achieved up until now. Near real time refers to the rate of detection of the relevant quantities of nuclear materials during the various stages of the SNF reprocessing in the span of a few hours, which is generally commensurate with the rate at which the SNF is reprocessed. While techniques for measuring near real time in bulk quantities, e.g., the volume of dissolved fuel and flow rates have been developed, Near Real Time Accountability related to on-line measurement of the elemental and isotopic concentrations was not possible with conventional detection methods (e.g., with K-Edge densitometry; X-ray fluorescence; Hybrid K-Edge/X-ray fluorescence densitometry; mass spectrometry; high resolution gamma spectrometry; isotope dilution gamma spectroscopy; constant coulomb coulometry; titrimetry; gravimetry; spectrophometry; calorimetry). In general, current methods do not offer Near Real Time Accountability capability for isotopic assessments, but rather require off-site shipment to special laboratories for relevant quantities data which may take from a few days to weeks. More importantly, current methods do not provide the means to determine Special Nuclear Materials isotopic inventories in-situ.

A detection system and methodology that permits on-line assessment of the U and Pu type actinides at the earliest stages (and subsequent stages), and that is complemented with current technology (e.g., simple methods of measuring weight and volume) for later stages would provide major improvement to reprocessing both in operational efficacy and in safety. A framework and methodology that achieves this goal by using a Tension Metastable Fluid Detector (TMFD) sensor system is described.

Spent Nuclear Fuel from a typical light water reactor contains a large collection of fission products with isotopes that span the periodic table from ⁷²Fe to ¹⁶⁷Er (plus a minor amount of tritium from tertiary fissions). In addition, Spent Nuclear Fuel contains radioactive activation products and transuranic actinide elements (i.e., Pu, Np, Am and Cm). While the majority of the fission products are gamma-beta emitters, it is the alpha-emitting uranium and transuranic isotopes that cause significant security as well as health safety concerns. Table 1 depicts the inventory of uranium and transuranic elements in representative spent fuel assemblies from a pressurized water reactor. All of the uranium and transuranic isotopes emit alpha particles but only some of them generate also significant quantities of neutrons from spontaneous fission.

The data in Table 1 illustrate both common grounds and differences between U and Pu isotopes in Spent Nuclear Fuel. The commonality is that both isotope groups exhibit alpha particle emissions with energies defined by the individual isotopes that vary between about 4 MeV to about 6 MeV. The differences arise in neutron emissions due to spontaneous fission. Uranium has a maximum emission rate in Table 1 of about 10⁴ n/s/MTU for ²³⁸U. Such value, when diluted and spread out over space in piping (e.g., to over 1 m² of surface area), may be difficult to passively measure over cosmic background neutron fluxes. On the other hand, some of the transuranic isotopes like ²⁴⁴Cm can emit about 10⁹ n/s/MTU. Such emission constitutes a readily measurable quantity (diluted or otherwise), using Tension Metastable Fluid Detector technology even in extreme gamma-beta fields wherein conventional sensors are saturated.

TABLE 1 Composition of a typical SNF assembly from a 2000 MW_(t) PWR. (Takahama-3, Initial U-235 enrichment: 4.11%, Burn-up: 47.03 GWd/mtu, Cooling time: 0 y) Spontaneous Fission (SF) Alpha Alpha Neutron Mass Half Life Ratio energy Emission Isotope (kg · MTU) (years) (%) (KeV) (n/s) ²²⁹Th 7340 100 5167.6 ²³⁰Th 0 7.54 × 10⁴ 100 4770 ²³²Th 0  1.41 × 10¹⁰ 100 4082.8 ²³²U 68.9 100 5413.55 ²³³U 0 1.59 × 10⁵ 100 4908.6 ²³⁴U 0.187 2.46 × 10⁵ 100 4858.5 2.04 ²³⁵U 7.93 7.04 × 10⁸ 100 4678.8  1.19 × 10⁻¹ ²³⁶U 5.53 2.34 × 10⁷ 100 4572 3.59 × 10¹ ²³⁸U 925 4.47 × 10⁹ 100 4270 1.87 × 10⁴ ²³⁶Np 0 1.54 × 10⁵ 0.16 5007 ²³⁷Np 0 2.14 × 10⁶ 100 4959.1 ²³⁸Pu 0.319 87.7 100 5593.2 1.08 × 10⁶ ²³⁹Pu 5.98 24110 100 5244.5 1.28 × 10² ²⁴⁰Pu 265 6563 100 5255.78 4.08 × 10⁶ ²⁴¹Pu 1.75 14.35 0.00245 5140.1  3.86 × 10⁻⁶ ²⁴²Pu 0.834 3.73 × 10⁵ 100 4984.4 1.95 × 10⁶ ²⁴⁴Pu 0 8.08 × 10⁷ 99.879 4665.5 ²⁴¹Am 0.0533 432.2 100 5637.8 8.75 × 10¹ ²⁴²Am 0.0012 141 0.459 5588.34 6.30 × 10¹ ²⁴³Am 0.193 7370 100 5438.1 1.52 × 10² ²⁴²Cm 0.0162 180 days 100 6.12 1.14 × 10⁸ ²⁴³Cm   8 × 10⁻⁴ 29.1 99.71 6168.8 2.48 × 10² ²⁴⁴Cm 0.0882 18.1 100 5901.61 1.04 × 10⁹ ²⁴⁵Cm 0.00592 8500 100 5623.5 6.61 × 10² ²⁴⁶Cm 7.55 × 10⁻⁴ 4730 99.9737 5474.8 6.51 × 10⁶ ²⁴⁷Cm 1.07 × 10⁻⁵ 1.56 × 10⁷ 100 5353.3 ²⁴⁸Cm 0 3.40 × 10⁵ 91.61 5161.73 ²⁵⁰Cm 0 9000 8 5169

By sampling on-line for characteristic neutron (including multiplicity) and alpha emission spectra, the presence or absence (via convolution) of the Pu versus U component can be readily confirmed amidst the mix of isotopes of Cm, Np and Am. Multiplicity refers to a key aspect of fission, that two or more neutrons are released virtually simultaneously, the feature which helps to distinguish neutrons emitted from the actinides which do have multiplicity from those neutrons, including background cosmic neutrons or neutrons released from non-fission processes, that lack multiplicity (i.e. neutron produced one at a time). The level of multiplicity for each actinide of interest is different and can be described by the following equation: Spontaneous Fission Multiplicity=0.27318Z−22.7734 wherein Z refers to the atomic number of the element. Using this formula, the multiplicity of the actinides of interest can be calculated: Th (Z=90)=1.81, U (Z=92)=2.36, Np (Z=94)=2.9, Am (Z=95)=3.18, Cm (Z=96)=3.45, Cf (Z=98)=3.99. By monitoring the number of neutrons released simultaneously (within no more than a few picoseconds) the respective actinide type can be identified. Unfortunately, known systems, require off-line chemical analyses or counting methods by taking samples to a testing laboratory. Reasonably accurate detection of U and Pu actinides in spent fuel compositions is complicated by the high beta-gamma radiation levels (about 10¹⁶ β or γ/s per assembly at about 1 year after shutdown) and the complexity associated with SNF composition.

In general, fluid metastable states can be reached via tensioning at ambient temperatures. Metastable states can also occur via thermal superheating at high positive pressures followed by depressurization such that fluids become sensitive to incoming radiation and form bubble tracks. When in a metastable state (either tensioned or superheated), explosive phase changes are triggered by stimuli that provide the excess energy required to reach the stability limit at which point the liquid changes phase. Stimuli may include extremely high nucleation rate-inducing nuclear particles such as neutrons, alphas, fission fragments, gamma photons as well as visible (collimated) photons from a laser. The thermodynamic phase spaces associated with tension and thermal superheat-based fluid metastability are depicted in the P-V diagram shown in FIG. 2. As the state of the fluid approaches the stability limits, as shown in FIG. 2 (a.k.a. the spinoidal limits of tension and thermal superheat respectively), the number of nuclei undergoing phase change starts to increase—reaching levels of about 10²⁵ nuclei/mL·s at the stability limits. As the tension or thermal superheating of the fluid moves away from the stability limits, the addition of excess energy becomes necessary for triggering phase change. Upon triggering of the phase change in metastable fluids, stored energy is released via vaporization growth of fast nucleating vapor bubbles. If the thermal energy deposition rate is sufficient to nucleate a critical size vapor nucleus, generally in the nanometer range, the nucleus will continue to grow into a macroscopic visible vapor bubble.

While bubble chambers and Superheated Drop Detectors (SDDs) operate in the positive pressure superheat regime, Tension Metastable Fluid Detector technology is distinct in its operation in the diametrically opposite regime (i.e., tensioned metastability without superheat). For any given tensioned metastable state far from the spinoidal limit, the required excess energy for triggering phase change of liquids and bubble formation must be provided by energetic ionizing particles such as neutrons, alphas, fission fragments, etc. For a given level of tension metastability, the excess energy required for forming bubbles will furthermore vary with the type and energy of radiation (i.e., neutrons vs alphas vs fission products vs photons), since it is well-known that the linear energy transfer (LET) or dE/dx is strongly dependent on the type of radiation involved, and this can be used to distinguish the types of radiation. This property, which enables macro-mechanical detection of nuclear-scale particles, can be used in ultra-sensitive detectors for nuclear engineering and science applications such as reactor power monitoring, identifying emissions from WMD-based special nuclear materials, or for online monitoring of nuclear spent fuel reprocessing streams where tons of Special Nuclear Materials are processed.

The Tension Metastable Fluid Detector sensor technology is based on placing ordinary fluids such as water or acetone in thermodynamic states of “tension” metastability under vacuum conditions (e.g., −5 bar) at room temperature.

Once the liquid molecule bonds are stretched, excess energy deposited from the direct strike of an energetic particle (e.g., a neutron or alpha particle with energies ranging from keV to MeV) onto a tensioned metastable fluid results in the nucleation of nanoscale bubbles which grow to visible size and then implode back to the liquid state accompanied by audible shock signals and light flashes which can be recorded using conventional electronics. The type and energy of the incident radiation can be assessed by monitoring the energy deposition rate (dE/dx) and the tensioned state and specific properties of the fluid in order to determine the individual actinide species in the mix of nuclear material.

SUMMARY OF INVENTION

Compositions and methods for monitoring the quantity of an actinide present in a test sample are disclosed. The method involves obtaining a test sample for which the knowledge of the actinide amounts is desired and obtaining a tensioned metastable fluid detector having a fluid and a fluid tension level such that the radioactive emission from the actinide can be detected and then determining the amount of the actinide in the sample.

Compositions and methods for monitoring the motion of special nuclear materials through space are also described. The method involves obtaining an acoustically tensioned metastable fluid detector having a fluid and a tension level such that a special nuclear material can be detected and monitoring the direction of the special nuclear material at least at two different times. The difference in the location of the special nuclear material as a function of time then provides an indication of the motion of the special nuclear material through space.

Compositions and methods for monitoring the quantity of a fissile special nuclear material present in a test sample are disclosed. The method involves obtaining a test sample that may contain a fissile special nuclear material. The emission from the sample is then measured using a tensioned metastable fluid detector having a fluid and a fluid tension level such that the fissile special nuclear material can be detected.

Compositions and methods for monitoring actinides during reprocessing of spent nuclear fuel are disclosed. In certain methods the fuel can be monitored after a typical 30-year cool down or waiting period according to Algorithm 1 below. In Algorithm 1 the method for real-time passive monitoring of actinides can include the steps of:

-   -   1. procuring a spent nuclear fuel sample of known initial         enrichment (i.e. known fraction of ²³⁵U isotope in the uranium         fuel mixture);     -   2. estimating the amount actinides using a computer program,         such as Origen;     -   3. determining actinide masses and spent fuel neutron production         (α, n) production;     -   4. calculating a activity of actinides using masses from ORIGEN         and Half-life (T½);     -   5. measuring neutron production with a TMFD system (Calibrated         with ²⁵²Cf and PuBe);     -   6. comparing in real time, the measured neutron production with         the predicted neutron production, and if the calculated and         predicted neutron production numbers do not agree, re-calibrate         the Origen program code input model (e.g., for core-averaged         fuel burn up) such that the numbers do agree;     -   7. extracting a sample from the dissolved fuel and dilute (e.g.,         to about −0.2 decay per second if near real-time detection is         desired to stay within about 5 seconds (=1/0.2)) based on ORIGEN         activity estimates of alpha, spontaneous fission and neutron         emission;     -   8. confirming relative absence of ²⁴²Cm activity (i.e. for alpha         activity);     -   9. measuring ²⁴⁴Cf and determine the concentration of ²⁴⁴Cm in         the sample;     -   10. measuring combined ²³⁸Pu, ²⁴¹Am, and ²⁴⁴Cm and determine the         concentration of ²³⁸Pu, ²⁴¹Am in the sample;     -   11. re-calibrating and refine ORIGEN-S Model for consistency         with experimental findings on Cm, Am, and Pu;     -   12. determining the concentrations of ²³⁹Pu from ²⁴¹Am, ²³⁸Pu,         ²⁴⁴Cm as well as ORIGEN code ratios;     -   13. cross verifying the determined concentrations of ²³⁹Pu from         extraction stream by active CTMFD monitoring or CTMFD sipping         based monitoring of ²³⁸Pu and ²³⁹Pu.

Compositions and methods for monitoring actinides during reprocessing of spent nuclear fuel after a typical short term cool down period of 180 days are disclosed below in Algorithm 2. In Algorithm 2 the method for real-time passive monitoring can include the following steps:

-   -   1. enriching a spent nuclear fuel sample;     -   2. estimating the amount actinides using a computer program such         as ORIGEN-S     -   3. determine actinide masses and spent fuel neutron production         (α, n) production, calculate α activity of actinides and T½;     -   4. measuring neutron production rate and multiplicity (from         spontaneous fission) with TMFD system calibrated with ²⁵²Cf and         PuBe, compare measured neutron production with predicted neutron         production;     -   5. if the calculated and predicted neutron production numbers do         not agree re-calibrate the ORIGEN input model such that the         numbers do agree;     -   6. extracting a sample from the dissolved fuel and dilute (e.g.,         to about 0.1 to about 10 decays per second, and more preferably         about 0.2 decay per second if near real time detection is         desired to within about 5 (=1/0.2) seconds) based on ORIGEN         activity estimates of alpha, spontaneous fission and neutron         emission; then, confirm relative absence of ²⁴²Cm activity (i.e.         look for alpha activity), however any detection time can be used         for example 1 to 60 or 100 seconds could be used;     -   7. measuring ²⁴²Cm and determine the concentration of ²⁴²Cm in         the sample;     -   8. measuring combined ²⁴⁴Cm and ²⁴²Cm and determine the         concentration of ²⁴⁴Cm in the sample;     -   9. measuring combined ²³⁸Pu, ²⁴²Cm and ²⁴⁴Cm and determine the         concentration of ²³⁸Pu, in the sample;     -   10. measuring combined ²³⁸Pu, ²⁴¹Am, ²⁴⁴Cm and ²⁴²Cm and         determine the concentration of ²⁴¹Am in the sample;     -   11. re-calibrating and refine ORIGEN-S Model for consistency         with experimental findings on Cm, Am, and Pu;     -   12. determine ²³⁹Pu from ²³⁸Pu, ²⁴¹Am, ²⁴⁴Cm and ²⁴²Cm as well         as ORIGEN code ratios;     -   13. cross-verifying the determined concentrations of ³²⁹Pu from         ²⁴¹Am, ²³⁸Pu, ²⁴⁴Cm with downstream levels of ²³⁹Pu from Pu         extraction stream by active CTMFD monitoring or CTMFD sipping         based monitoring of ²³⁸Pu and ²³⁹Pu.

DESCRIPTION OF FIGURES

FIG. 1 provides a schematic diagram of a high-level process flow diagram for the UREX+ aqueous separations plan. This is illustrative of the general flow of events during reprocessing; other schemes such as PuREX are similar. The methods disclosed are applicable to any such reprocessing scheme.

FIG. 2 provides a diagram of showing thermodynamic phase-space for tension and superheated fluid states.

FIG. 3 provides schematic diagrams of Tension Metastable Fluid Detector systems where the metastable tension in the fluid is induced via: (a) induced oscillating pressure fields and (b) centrifugal motion.

FIG. 4 shows the results of alpha spectroscopy with Tension Metastable Fluid Detectors with NIST-certified actinides (Note: Loess is a smoothing algorithm)

FIG. 5 provides a graphical representation of Neutron flux ratio versus angular direction for a PU-Be source.

FIG. 6 provides graphical representations of detection time versus P_(neg) (detector fluid tension pressures) for a CTMFD configuration showing the ability to discriminate between neutron energies and emission spectra.

FIG. 7 graphically depicts tension thresholds for Pu, U, Cm and Am isotopes using a single Tension Metastable Fluid Detector system.

FIG. 8 provides a graphical depiction showing the relationship between the average signal waiting time in a Tension Metastable Fluid Detector and a fluid tension for various actinides.

FIG. 9 depicts the relationship between a decay/sec/MTU and burn up (GWd/MTU) for various actinides over a 180 day cooling period.

FIG. 10 provides a bar graph showing the variation between neutron yield isotope content from spontaneous fission and α-n reactions: 4 without ²³⁵U; 40 GWd/MTU burn up; 6 month (180 day) Spent Nuclear Fuel cooling period.

FIG. 11 graphically depicts the relationship between α yield and burn up over a 30 year cooling period for various elements in spent nuclear fuel.

FIG. 12 graphically depicts the relationship between neutron yield with isotopic content from spontaneous fission and α-n reactions: 3 without ²³⁵U; 30 GWd/MTU burn up over a 30 year Spent Nuclear Fuel cooling period.

FIG. 13 provides an illustration of a detection system for detecting neutrons from a sample having five TMFD detectors set to distinct P_(neg) values.

DETAILED DESCRIPTION OF INVENTION

Oak Ridge Isotope Generation (ORIGEN) code, developed by the U.S Department of Energy's Oak Ridge National Laboratory is a nuclear fuel depletion analysis program which was originally developed to monitor for fuel depletion from fission as well as the various fission products over time, and then to monitor for various nuclear reactions and radioactive decay chains which follow. Various versions of ORIGEN have been developed, and when it is used as part of another suite of codes, a suffix is attached, e.g., ORIGEN-S (the “S” depicting the so-called SCALE code package). In this invention, the terms ORIGEN and ORIGEN-S denote the same underlying computer code model.

A composition is disclosed that provides a simple low cost class of sensors with high intrinsic efficiency, for example about 90% or more efficiency, that are able to distinguish between neutrons, alpha particles, and fission fragments and simultaneously also provide directionality and multiplicity related information for neutron emissions from a single, portable sensor system for which the detection efficiency can be controlled. These detectors can physically “see” and “hear” radiation while also deriving spectroscopic information and discerning the direction of incoming radiation and at the same time remain “blind” to gamma photons. The spectroscopic information is acquired using TMFDs by sweeping the tension pressure states (P_(neg)) across a range of pressures. At certain pressure points detection of various energy ionizing particles is possible. The ability to remain blind to gamma photons and beta particles allows for use of the device in the intense radiation fields of spent nuclear fuel to decipher neutron and alpha emissions characteristic of U and Transuranic isotopes. Tension Metastable Fluid Detectors remain gamma blind even in the intense field of an operating 1,000 MWe power reactor. Table 2 summarizes certain characteristics of Tension Metastable Fluid Detectors.

TABLE 2 Comparison of Tension Metastable Fluid Detectors vs State-of-Art Systems Parameter State-of-Art Systems TMFD System Size, standoff Limited to small sizes Can be tailored to situation (single Size: physical (cost exponentially large system) dimensions of detector; increases with size) Standoff: how far the detector is from the source of radiation Intrinsic efficiency: ~10-20% (MeV neutrons); ~90% (MeV neutrons and thermal (The fraction of ~90% (thermal neutrons neutrons). incident radiation for 30 cm × 30 cm) passing through the detector that is actually detected) On-Off times Large (minutes) Microseconds (How long it takes for the detector system to switch on and off) Gamma blind? No; systems can get Yes; No saturation problems. saturated in high gamma fields Directionality/Direct No. Yes (to within 10° for Source Imaging: directionality; also, for (The ability to Special Nuclear Materials characterize the shape source imaging and size - like taking a picture) Cost High (>$10K for simplest Low-to-modest ($0.1K to $1K). systems). Complexity Large; requires complex Low; Can actually see and hear electronics. radiation in the form of collapsing bubbles. Same system for No. Require specialized Yes. Same system can be tailored neutrons, photons, systems for each particle to detect neutrons, photons and alphas? type. alphas. Multiplicity with No. Requires multiple Yes (also aids in identifying single detector systems and complex Special Nuclear Materials as system? electronics. opposed to cosmic and other background radiation).

Multiplicity as it pertains to neutron emission comprises two or more simultaneous (i.e., occurring so fast within pico to femto seconds that for practical purposes are deemed to be simultaneous) neutron emissions which occur only during fission events as compared with randomly generated neutrons from either radioactive decay or from non-fission nuclear reactions (like alphas striking nuclei of elements like Be, B, Li, F, O). Fission of U, Pu and other fissile elements produces different numbers of neutrons in each fission event. Hence, this method provides only detection of fissile materials but also identification of the specific element type. Neutron multiplicity capability for a detector is the ability to detect two or more neutrons arriving into the detector simultaneously. In ordinary detectors which are enclosed and based on counting scintillation light pulses or charge pulses and without the enablement of monitoring of the individual strikes, such events get detected as being from a single event. In the TMFD system, the occurrence of two or more simultaneous bubble formations within the TMFD volume becomes conspicuous and can be recorded using a single imaging and/or electronic recording system. In this regard, we have, as noted earlier, provided for the multiplicity values for various actinides of interest ranging from Th to Cf. A single TMFD can detect and reveal arrival of simultaneous neutron emission. This makes the TMFD far more efficient for deciphering the specific element compared with conventional detectors.

The multiplicity (for spontaneous fission) varies from less than 2 to close to 4 for Cm and Cf. The multiplicity (ν(E_(n))) value can be increased by increasing the energy of external neutrons (E_(n)) used for interrogating a given mass of fissile materials. In this case, (ν(E_(n))=ν(0)+aE_(n) where, a varies with the nuclide in question [i.e., a=0.1419 (²³⁵U); =0.1482 (U-238); =0.1432(²³⁷Np); =0.1471 (²³⁹Pu); =0.1482(²⁴¹Am); =0.1536 (²⁴⁴Cm)]. For example, one usually has available portable D-T accelerators which produce 14 MeV neutrons. If we use 14 MeV neutrons to decipher the presence of ²³⁵U, for example, the value for ν(14) for ²³⁵U induced fast neutron fission would equate to about 4.5 (versus about 2.36 for spontaneous fission); similarly, with 14 MeV neutrons, the induced fission with ²³⁹Pu would move up from about 2.9 (spontaneous fission) to about 5. Multiplicity-based determination can be accomplished more efficiently using multiple TMFDs surrounding the interrogated item to increase the solid angle subtended onto the source of materials.

FIG. 3 shows schematic diagrams for two types of Tension Metastable Fluid Detector systems that have been developed and qualified in laboratory experiments. The Acoustically Tensioned Metastable Fluid Detector (ATMFD) shown in FIG. 3a uses piezoelectric elements to induce time-varying acoustically-driven oscillating pressure fields (compression and tension) in a resonance mode at micro-second time scales much like a laser cavity. When in the tension mode, the fluid field nucleates bubbles in transient fashion when nuclear particles provide sufficient energy. As such, the ATMFD turns on and off within microseconds. Interestingly, while in compression mode the system remains completely blind to all forms of radiation which is important in pulsed (neutron/photofission) based monitoring. Conventional detectors saturate during the pulsing time span and continue to remain blind for considerable periods of time after pulsing; therefore, losing important emission signatures. The ATMFD's unique features allow it to overcome such dead-times. The location and timing of the bubbles provides information on type, multiplicity, energy and directionality of the nuclear radiation.

In another embodiment, tension metastable states in fluids are created with centrifugal force. These detectors are useful in centrifugal tension metastable fluid detector system. FIG. 3b of the centrifugal tension metastable fluid detector system depicts an enclosure constructed from glass tubing formed into a diamond shaped apparatus. The apparatus is partially filled with a working liquid of density ρ and meniscus separation 2r and attached to a variable speed motor. Upon rotation, centrifugal force pulls the molecules outward effectively placing the molecules in the central bulb region in a tensile state. The level of tension or negative pressure ρ_(neg) on the centerline is given by the following equation. ρ_(neg)=2×π² ×ρ×r ² ×f ²−ρ_(amb) where, f is the rotational frequency and ρ_(amb) is the ambient pressure. As a first order approximation, the pressure variation in the central bulb region can be modeled as flow between two cylinders rotating with the same velocity where the inner cylinder has a radius of zero. This approximation reduces the equation to the Bernoulli equation. For the small bulb radii used in CTMFD apparatus the pressure variation in the central bulb region is negligible. Both system designs are amenable to scalability to enhance overall efficiency and sensitivity.

The Tension Metastable Fluid Detector systems can be used to monitor trace, such as sub-picoCurie/mL, actinide quantities via direct sampling in real-time and with spectroscopic information at levels about 100 times below the resolution of liquid scintillation spectrometry. This is clearly shown in FIG. 4, as example of the detection within about 35 s of U isotopes in the 0.015 Bq/g (equal to about 0.4 pCi/mL) range. The CTMFD volume was about 2 cc. This activity level was found to be below electronic background noise and indistinguishable when tested in a LS6500 Beckman spectrometer. Also, by merely increasing the size of the central CTMFD bulb by a factor of 10, e.g., from about 2 cc to about 20 cc, one could readily get similar turnaround results in about 35 s for 10-fold lower activity, e.g., even at 0.04 pCi/mL levels. The data points shown in FIG. 4 represent averaged values. Radioactivity, being a random process, necessarily implies a Poisson distribution; e.g., at about 35 s wait time, the spread can be expected to be about +/−6 s representing about 70% of all measured counts for 1 standard deviation. In this regard, a Tension Metastable Fluid Detector system can be used to distinguish between ²⁴¹Am and ²³⁸Pu alpha recoil emissions, which are only about 2 keV apart, and the same system can also detect spontaneous fission events.

FIG. 4a illustrates the detection of trace actinide bearing samples having only about 0.05 Bq/cc. Special Nuclear Materials that include actinides ranging from ²³⁸Pu, ²³⁹Pu, ²⁴¹Am, ²³⁴U, and ²³⁸U can be detected and monitored by tailoring the specific level of tension metastability. These data were obtained using NIST-calibrated sources. Separately, FIG. 4b illustrates that the about 2 keV separation between actinide recoil energies can be observed for trace-level isotopes virtually in real time.

A method is also disclosed for monitoring neutron emissions with greater than 90% intrinsic efficiency, and with ATMFDs for discerning the direction of a Pu—Be neutron source (±about 30°) with 90% (FIG. 5). This evidence formed the basis for extending the technology not only for real-time neutron source directionality, but also for simultaneous source imaging such that the actual motion through space of Special Nuclear Materials can be monitored and tracked.

Monitoring can also distinguish between fission-induced neutron multiplicity and random neutron events. This surprising possibility was observed where 8-fold greater multiple neutron-induced events were recorded when using a relatively weak spontaneous fission source (i.e., ²⁵²Cf) of about 10⁵ n/s strength, in contract to when a Pu—Be random neutron emitting (about 10⁶ n/s) source was used. This provides a basis for discerning between fissile Special Nuclear Materials (U to Pu to Cm) from their multiplicity signature differences and rejecting extraneous random events (e.g., the well-known “Ship-Effect”). As noted earlier, fission events lead to neutron multiplicity, whereas, neutron emissions from radioactive decay and non-fission nuclear reactions lead to randomly produced non-simultaneous neutron emissions.

The Tension Metastable Fluid Detector is blind to gamma radiation while detecting neutrons and alpha radiation for fields greater than about 10¹¹ γ/s, which is equivalent to the gamma field about 5 m away from a spent fuel assembly after about 6 months of cooling. It has been estimated that Tension Metastable Fluid Detectors that are tailored for alpha, neutron, or fission fragment detection can remain blind to energetic gamma photons even within the core of an operating 3,000 MW(t) nuclear reactor.

The gamma flux in a 3,000 MW(t) nuclear fission reactor is well-known to be in the range of about 10¹⁴ γ/cm²/s. The generation of detectable bubble events in TMFDs requires a certain threshold level of energy in the range of about 100 keV deposited by a recoiling carbon or oxygen type nuclei within the dimensions of the critical radius of about 70 nm. The maximum energy produced from a typical 1 MeV nuclear reactor gamma photon on to nuclei such as H and C can at most be 0.5 keV per collision. Studies based on pulsed nanosecond lasers and theoretical assessments indicate that gamma photon influence on to TMFDs if placed within an operating power reactor could only take place if the power level and hence, the photon flux, were to be over 10²³ γ/cm²/s—which is a billion times greater than in existing nuclear reactors.

A Tension Metastable Fluid Detector systems for Near Real Time Accountability monitoring of key Special Nuclear Materials for Pu, in particular for ²³⁹Pu; other uranium species, especially ²³⁵U; and Cm actinides in various sections of a reprocessing plant are disclosed.

The principal isotopes of interest for security purposes are ²³⁹Pu and ²³⁵U, both fissile isotopes. While both of these isotopes are abundant in mass in Spent Nuclear Fuel, neither their alpha nor spontaneous fission activity levels are high enough (relative to the background radiation levels in Spent Nuclear Fuel) to be readily detectable. The high background alpha and neutron emissions in Spent Nuclear Fuels arise principally from the formation of ²⁴²Cm, ²⁴⁴Cm, ²⁴¹Am and ²³⁸Pu. The level of background from these isotopes are, in general, at least an order of magnitude greater than the alpha or neutron activity of ²³⁹Pu, and several orders of magnitude greater than that of ²³⁵U. This could readily be overcome by resorting to active interrogation using an external neutron source because the fission cross-section of ²³⁹Pu is large, for example greater than about 600 barns, and the mass quantity of ²³⁹Pu is orders of magnitude greater than that for the Cm and Am isotopes. However, such a procedure, although enabling and possible to undertake, requires use of an external neutron or photon source such as a D-D or D-T accelerator for neutrons, or an electron linear accelerator (LINAC) for photons, or, use of isotope neutron sources such as ²⁵²Cf or Pu—Be or Am—Be “together” with TMFD banks. This can raise the overall system costs and complexity (e.g., accelerator systems can cost upwards of $100K to $1 M) and hence, would not be as economic or portable (due to their weight and fragile electronic components) as using simple passive means which relies mainly on TMFDs and PC-based processing algorithms alone. Therefore, in the absence of active monitoring, e.g., neutron or photon-based fission of the target substance, indirect quantification of the amounts must be used. This is especially relevant for ²³⁹Pu, the Pu isotope of greatest interest found in a nuclear explosive device Importantly, the International Atomic Energy Agency (IAEA) has set the threshold limit for the “significant quantity” of Pu (including all isotopes) at just 8 kg. The quantity of ²³⁹Pu in Spent Nuclear Fuel is difficult to determine in the initial reprocessing steps because ²³⁹Pu is mixed with extremely high levels of fission products. The high levels of beta-gamma activity in SNF, including radiation intensity fields of over 100 R/h, make it virtually impossible for present-day sensors (e.g., ³He detectors) to provide meaningful information on actinide content in general, much less ²³⁹Pu levels.

A Tension Metastable Fluid Detector system can be used to monitor the collection of actinides (including ²³⁹Pu) at the highly-sensitive front-end of the PuREX/UREX reprocessing streams since it is gamma-beta blind, while remaining selectively sensitive with over about 90% efficiency for detecting alpha recoils, neutrons and fission fragments from actinides.

The described methods rely in part on the following assumptions:

-   -   The original ²³⁵U enrichment in the Spent Nuclear Fuel (i.e.,         prior to fission) is known. This value is readily and         contractually available to the nuclear power utility from the         fuel vendor and hence, is a known quantity with little to nil         uncertainty.     -   The power history of the fuel assembly is known as it was         operated in the reactor over a given period of time while         producing power. These data are often preprogrammed by the         utility during the development of core management schemes, and         records are kept for control rod motion and sensors during any         particular cycle. This parameter influences the degree of fuel         burn up throughout the core. Although the power history in a 3-D         sense is reasonably known, unforeseen circumstances such as         reactor scrams or other aspects requiring temporary reactor         shutdown can give rise to variations. Hence, a margin of         uncertainty results—the degree of which must be quantified via         actual measurements for key actinides (e.g., ²⁴²Cm, ²⁴⁴Cm,         ²⁴¹Am, ²³⁸Pu, ²³⁵U, ²³⁹Pu) in order to then correct for the         averaged burn up for a given SNF assembly or assemblies which         are dissolved in a vat ahead of reprocessing.     -   The cool down period of the Spent Nuclear Fuel is known and         available from the data logs kept by the nuclear power utility.         This parameter is known with good confidence and hence, with         minimal uncertainty.

The ORIGEN-S depletion code is available to simulate (with reasonable accuracy) the burn up history and buildup of actinides and fission products. ORIGEN-S is a computer code widely utilized worldwide and available from the U.S. Department of Energy (USDoE's) Oak Ridge National Laboratory, Oak Ridge, Tenn., USA. ORIGEN-S has been validated extensively and also separately against data for this application as discussed subsequently.

As part of the real-time on-line monitoring with Tension Metastable Fluid Detector Systems, the validity of the ORIGEN-S code as a virtual simulator to provide a first-cut estimate of Special Nuclear Materials actinide content in Spent Nuclear Fuels was tested. It was useful to assess how well predictions compare with reasonably well characterized post-irradiation-examination data. Argonne National Laboratory and Pacific National Laboratory and Oak Ridge National Laboratory staff have conducted assessments for such situations and, post-irradiation-examination on several light-water-reactor Spent Nuclear Fuels in 2007, referring to these samples as Approved Testing Materials (ATM) as part of a DoE program for developing experimental material for nuclear waste repository researchers. Utilizing the information on power history, initial enrichment and cool down histories, ORIGEN-S based models were developed for predicting fuel depletion and the generation of key actinide inventories over time. A sample of comparison against ATM-103 Spent Nuclear Fuel specimen is shown in Table 3. As noted in Table 3, the ratio of ORIGEN-S to post-irradiation-examination values is within +/−8% for the mix of actinides; importantly, for ²³⁹Pu and ²³⁵U the comparison is within 3% to 1%, respectively.

TABLE 3 Comparison of Predicted (ORIGEN-S) and PIE Data - PWR Fuel burn up for a 30 MWd/MTU (average); 2.72 wt. % enrichment; 6.5 y cooling time. ORIGEN PIE Nuclide (kg/MTU) (kg/MTU) ORIGEN/PIE ²⁴¹Am 0.377 0.382 0.99 ²³⁷Np 0.404 0.373 1.08 ²³⁸Pu 0.157 0.168 0.93 ²³⁹Pu 4.9 4.75 1.03 ²⁴⁰Pu 2.42 2.4 1.01 ²⁴¹Pu 0.901 0.922 0.98 ²⁴²Pu 0.594 0.621 0.96 ²³⁴U 0.143 0.136 1.05 ²³⁵U 5.38 5.42 0.99 ²³⁶U 3.63 N/A N/A ²³⁸U 947 955 0.99

The above comparisons show that, if the detailed power history, initial enrichment and cool down history are known with good confidence (e.g., core averaged burn up of fuel to within +/5%) for each Spent Nuclear Fuel, a reasonable estimate of ²³⁹Pu and ²³⁵U actinides (with over 95% confidence) can be made. However, detailed accurate information may not always be available, and even a seemingly small deviation of 3% of mass inventory from a total annual inventory of 1,000 kg could amount to about 30 kg or more for ²³⁹Pu which significantly exceeds the IAEA safeguards limit of 8 kg. Therefore, based on the results of ORIGEN-S validation studies, estimated predictions of inventory of various actinides should only be used as a simulation tool as part of a mix, to arrive at a first estimate. The mix refers to a combination of prediction and in-situ real time data acquisition via TMFD data for confirmation and refinement of the predictive tool as described above in Algorithms 1 and 2 to arrive at a first estimate. But, for on-line monitoring in real-time, the threat of potential diversion requires a real-time verification-correction tool that offers a means to continually cross-check and update to refine the primary assumptions used to make ORIGENS-based predictions. Tension Metastable Fluid Detector technology can be used for this purpose, in tandem with ORIGEN-based predictions.

From a practical viewpoint, there are two types of Spent Nuclear Fuels for reprocessing each requiring a somewhat distinct algorithm of steps for discerning the key ²³⁵U and ²³⁹Pu isotopes as described in the Algorithms. These include: (1) legacy fuel with 2-3 wt. % enrichment with 20-30 GWd/MTU burn up followed by a 30-year cool down, and (2) more modern fuels with a 4-5 wt. % enrichment with burn ups up to 50 GWd/MTU followed by a 0.5-year cool down period. The significance of these differences in light of relative actinide buildup is in terms of buildup of ²⁴¹Am, ²⁴²Cm and ²³⁸Pu (both strong alpha emitters but weak spontaneous fission neutron emitters), and ²⁴⁴Cm (a strong alpha and spontaneous fission neutron emitter).

For the first Spent Nuclear Fuel type with 30-year cool down, the relative activity of ²⁴¹Am, ²⁴⁴Cm and a ²³⁸Pu far outpaces the strength of ²⁴²Cm, whereas for the second Spent Nuclear Fuel type with only a 0.5-year cool down period, the relative buildup of ²⁴¹Am is negligible and the dominant alpha-neutron activity is from ²⁴²Cm, ²⁴⁴Cm and ²³⁸Pu.

The significant differences in the actinide buildup of the two types of Spent Nuclear Fuel, as mentioned above, demand unique monitoring strategies depending on the fuel type. The two types of Spent Nuclear Fuel, however, will also possess certain commonalities, which are listed in this section ahead of two specifically targeted algorithm-based methods targeted to Near Real Time Accountability at the front-end. The two algorithms are described above. Common features between the two monitoring schemes are presented (along with estimated time of task completion).

First, information obtained from the nuclear utility is introduced into ORIGEN and is used to develop an estimate of the relative quantities of different actinides, including estimates for a range of potential burn up levels. This should only require a few minutes to accomplish using for example, a personal computer based system.

Second, Tension Metastable Fluid Detector systems are used to monitor the Spent Nuclear Fuel at the initial stages of reprocessing to determine the quantity of ²³⁹Pu. The amount of ²³⁹Pu can be determined based on monitoring of neutron activity in the presence of a very strong beta-gamma dominated radiation background that can be as high as about 10⁹ Ci (for a typical PWR using about 40 T of U and about 3-5 wt. % enrichment) in total at cycle end before cool down on a core average basis (comprising about 10²⁰ γ-β/s, about 10¹⁷ alpha/s, and about 10¹⁰ n/s). The SNF can be dissolved in nitric acid and placed in a vat prior to further reprocessing. The alpha particles cannot penetrate to the outside of the bath, but the neutrons, gammas and to a small extent the beta rays will penetrate. If a conventional detector is placed outside such a vat, the gamma to neutron flux would be about 10¹⁰:1 (i.e., over 10 billion times higher gammas compared with neutrons). Known detectors such as ²¹³Ne and ³He based detectors are limited in that they can reliably detect neutrons without gamma interference at most if the gamma to neutron fluxes are in the range of about 10:1 to about 10³:1. When these detectors are used in initial measurements of SNF a major uncertainty remains in terms of the quantity of actinides present, particularly ²³⁹Pu and ²³⁵U. Such uncertainty complicates monitoring the material as it passes through various processing stages. SNF from a typical PWR core at end of cycle can generate close to 500 kg of ²³⁹Pu, and substantially smaller quantities are considered a threat for the development of nuclear explosives. Moreover, at end of cycle not all of the ²³⁵U is consumed. About 500 kg of ²³⁵U remains which far exceeds what is considered a threatening level for nuclear explosives.

If knowledge about the quantities of actinide species can be made available right up front and throughout reprocessing procedures SNF could be handled more confidently and safely. A neutron detector that can detect neutrons with about 90% efficiency or more and that remains blind to gamma-beta radiation for use in monitoring SNF is disclosed. TMFDs offer such a capability. TMFDs using detection fluids such as acetone, isopentane, methanol, ethanol, trimethyl borate, perfluoroctane, R-113 and operated with P_(neg) down to about −20 bar have been demonstrated to remain totally gamma blind. They have also been observed to have over 90% of the theoretically attainable intrinsic efficiency for neutron, alpha and fission product detection. As a result the systems and methods described herein can be used to determine the quantity of ²³⁹Pu and other actinides present in SNF.

For SNF that is delivered to the reprocessing plant, the isotopic inventory which dominates the spontaneous fission neutron rate is ²⁴⁴Cm with an emission intensity of about 5×10⁸ n/s/MTU (0.5-year cool down fuel) and about 10⁸ n/s/MTU for 30-y cooldown fuel. For the Spent Nuclear Fuel the resultant neutron output will be about 3% to 10% greater due to the additional ^(244/242)Cm(α,n)¹⁶O reactions from the fuel being in UO₂ oxide form, but the ORIGEN assessment includes this factor resulting in approximately an additional 3×10⁷ n/s (0.5y cool down fuel) and about a 3×10⁶ n/s (30y cool down fuel). This addition of an (α,n) source to the spontaneous fission neutron source gives rise to a neutron spectrum that is a combination of two weighted spectra and can be readily accommodated. In this step, a pre-calibrated centrifugal tension metastable fluid detector (with a commercially available ²⁵²Cf, spontaneous fission source of certified intensity, together with a commercially available PuBe or AmBe type source of about 3 to about 10% of the ²⁵²Cf neutron intensity can be utilized at various distances from the pipe or vat holding the Spent Nuclear Fuel. Such a step provides the first sensor-based data for the presence of ^(244/242)Cm to update the ORIGEN simulation. This estimate may be further refined by extracting a small quantity of dissolved SNF and placing it in a TMFD system with a fluid such as acetone as the detection fluid and assessing for fission activity from spontaneous fission. The amount of extraction will depend upon the degree of dilution of the SNF. For 30y cool down fuel with about a 3 wt. % enrichment and 30 GWd/MTU burn up for example, the neutron production rate per MTU is estimated as: ²⁴¹Am about 1×10³ n/s; ²⁴²Cm about 3×10⁴; ²⁴⁴Cm about 8×10⁷ n/s; ²³⁸Pu about 2.6×10⁵ n/s; ²³⁹Pu about 80 n/s; and ²⁴⁰Pu about 2×10⁶ n/s. Clearly, ²⁴⁴Cm dominates in fission activity with all else being negligible by comparison. This means that the detection of ²⁴⁴Cm fission activity can be used to determine the relative quantities of other actinides as well. This can be readily determined by diluting the extracted fluid from the vat such that the expected fission activity is about 100 fissions/second based on the previously detected neutron activity as a whole. As shown in FIG. 6, the fact that the alpha activity for this sample will be several orders of magnitude higher does not interfere with this determination when the P_(neg) value of the TMFD is set such that the TMFD is sensitive only to fission events and blind to neutrons as well as alpha events. This can be seen in FIG. 6. By placing about 0.1% by volume of this actinide bearing fluid into 100 cc of acetone the resulting solution will have a “fission” activity of about 0.1 Bq/cc. This solution can be placed in a 1 cc TMFD at a P_(neg) of about −1 bar, and the TMFD will detect each fission event while remaining blind to alpha, neutron, gamma as well as beta activity. This measurement can be used to determine actual actinide activity levels together with the neutron measurements. For 180 day cool down fuel (40 GWd/MTU, 4 wt. % enrichment) the neutron activity from spontaneous fission is dominated by ²⁴⁴Cm (about 5×10⁸ n/s/MTU) and also ²⁴²Cm (about 1.3×10⁸ n/s/MTU) with the output from other actinides about similar to that for 30y cool down fuel except for ²⁴¹Am which is negligible. In this instance, the dilution steps would be similar to that taken for 30y cool down fuel followed with placing the TMFD system to a P_(neg) of about −1 bar to count for fission activity from Cm. Unlike for 30y cool down fuel, this measurement provides for the combined activity of ²⁴²Cm and ²⁴⁴Cm even though ²⁴⁴Cm activity for fission is about 5-fold higher for fission activity (not alpha activity). For 0.5y cool down fuel, the alpha activity is in fact dominated by ²⁴²Cm (about 1×10¹⁵ alphas/s/MTU) whereby, the ²⁴²Cm:²⁴⁴Cm alpha activities are in the ratio of about 5:1. This too could be readily ascertained. However, the same solution prepared for fission rate estimates would need to be diluted by a further factor of about 10 million times. For such a diluted system, alpha activity would now dominate and the P_(neg) can be swept from about −6 bar through −8 bar to then decipher for the activities first, for ²⁴²Cm and then, for ²⁴⁴Cm, respectively. This step may take approximately one to ten minutes to accomplish in a practical situation once the calibrated TMFDs are in place and tied in with the ORIGEN simulation platform.

Information from the detector system is used to make fission and neutron measurements which are then compared with the ORIGEN-predicted buildup of ^(244/242)Cm (the major source of neutron emission); in case of discrepancy, the Spent Nuclear Fuel averaged burn up would be adjusted such that the updated ORIGEN prediction for ^(244/242)Cm is commensurate with the measured value. Since the fundamental nuclear physics governing the burn up process builds up the other actinides in specific consort with ²⁴⁴Cm, this raises the confidence level of a best-estimate up front for all other actinides of interest, ²⁴¹Am, ^(244/242)Cm, ²³⁸Pu and, especially for ²³⁹Pu. Information from this step also provides the level of dilution of the actinide-rich fluid stream that will be necessary to dissolve within the working fluid of the Centrifugal Tension Metastable Fluid Detector for monitoring alpha activity from the various actinides. This can be done using a computing device.

The next measurement will involve directly measuring alpha activity. This involves removing a small quantity of SNF bearing reprocessing fluid. The amount of sample to remove is an amount that provides for suitable data acquisition times. The extent of dilution by the plant operator must also be known. Alpha activity in a given volume of reprocessing fluid is estimated on a per MTU basis. One MTU (in oxide form as is the case for virtually all nuclear power reactors) assumes a volume of about 10 L. Assuming a 10:1 dilution the resultant initial alpha activity would be: about 2×10¹⁰ Bq/cc (0.5y cool down fuel) and, about 2×10⁹ Bq/cc (30y cool down fuel). If detection of activity is to be about 10 s on average, the activity within a TMFD, such as a CTMFD having a sensitive volume about 1 cc can be set to be about 0.1 Bq (total activity). This means, the reprocessing fluid stream activity must be diluted (e.g., by over 2×10¹¹ times for 0.5y cool down fuel and by about 2×10¹⁰ times for 30y cool down fuel) to bring down the resultant activity in the TMFD fluid to about 0.1 Bq/cc (of TMFD fluid). For example, an aliquot (e.g., 1 μL) can be removed from the vat holding the dissolved Spent Nuclear Fuel, and diluted with acetone (as was done previously with NIST-certified standards) or with other suitable TMFD fluids such as ethanol, and methanol. The degree of dilution can be directly estimated based on the expected total activity such that the overall activity after dilution is in the 0.1 Bq/cc range for this example. Since the dilution is performed using the TMFD detection fluid (e.g., isopentane, acetone, methanol, or ethanol for example) the quantity of SNF reprocessing stream fluid in the TMFD fluid volume is negligible and far less than 1%. Levels of about 1% of nitric acid will not affect the TMFD detection. Assuming the centrifugal tension metastable fluid detector volume being used is 2 cc and the activity of the highest energy alpha emitting isotope ²⁴²Cm is 0.01 Bq/cc in the diluted solution, the time it will take to determine the presence of ²⁴⁴Cm would be about 50 seconds at a tension level of about −6.5 bar to −7.5 bar (per FIG. 7). If the activity of ²⁴⁴Cm is also 0.01 Bq/cc., then, the tension metastability level should be adjusted to sweep between −7 bar to −8.5 bar to detect the combined activity of ²⁴²Cm and ²⁴⁴Cm within 25 seconds. As shown in FIGS. 6-7 the negative pressure range can be scanned with a single centrifugal tension metastable fluid detector sensor to determine the amount of various actinides in a step-wise progressive fashion (i.e., first for ²⁴²Cm which emits alphas at 6.1 MeV; then for ²⁴⁴Cm which emits alphas at 5.8 MeV; then for ²³⁸Pu at 5.5 MeV; then for ²⁴¹Am at 5.49 MeV, and so on).

TMFDs can be calibrated for specific detection of the actinide species described above or other radiation sources. The specific values for P_(neg) associated with detecting the actinides (alpha emitting isotopes) shown in FIGS. 4-8 can be determined for any CTMFD or ATMFD system via calibration using NIST-certified sources. Small variations between TMFDs may occur and should be accounted for. Variations in a CTMFD can occur due to the central bulb not being axi-symmetric, as well as due to meniscus separation in the upper arms, the uncertainty in density due to temperature of operation, and other reasons. The meniscus separation, for example, involves taking an average of the outer and inner edges of the meniscus from each of the two arms. If the radius of separation is assumed to be from the outer edges, the resulting P_(neg) as computed (for a given density and rotational speed) will be greater than if the separation distance is assumed to between the inner edges of the meniscus. A consistent approach should be adopted to build up a calibration plot for time needed for detection of samples of NIST-certified (known) activity for specific actinide isotopes. These calibrations can easily be carried out by those of skill in the art.

Typically, the ²³⁸Pu:²³⁹Pu activity ratio is about 10:1. While ²³⁸Pu can be determined within about 10 s, to detect for ²³⁹Pu requires about 10-fold longer times of about 100 s; while still attainable directly, cosmic neutron induced background effects of about 0.0065 n/cm²/s should also be accounted for. The ²³⁵U activity in the overall process stream is normally expected to be much lower due to its half-life being about 1,000 times greater, although the total ²³⁵U mass at end of the cycle may be similar to that for ²³⁹Pu. This makes direct assessment for ²³⁵U in the overall process stream (upfront) somewhat impractical. While monitoring for ²³⁹Pu may be feasible, as mentioned above, the monitoring for ²³⁵U could only be carried out in the subsequent UREX stream (wherein, the U elements are preferentially diverted) and upon which higher alpha energy emitting elements of Cm, Am and Pu are absent. Overall, due to the significantly lower relative alpha activity of the ²³⁹Pu and ²³⁵U actinides, in order to monitor ²³⁹Pu and U-based isotopes directly, a centrifugal tension metastable fluid detector with a significantly larger sensitive volume of about 100 cc as shown in FIG. 3B must be used. In such an embodiment, the dimension of the radial separation term “2r” in FIG. 3B must be increased such that, variations of tension metastability within the central bulb are relatively small (e.g., about 1-5%) when compared with the overall variations between the central region and that at the end of the arms. In accordance with well-established laws of physics governing fuel burn up and isotope decay, and highly precise information relating to the dependence of ²³⁹Pu and U-isotope quantities on the quantities of ²⁴⁴Cm, ²⁴²Cm, ²⁴¹Am and ²³⁸Pu in the mixture, the levels of ²³⁹Pu and U-isotopes can also be determined. In order to expedite this, multiple CTMFDs working in parallel may be readily arranged in the system and used at P_(neg) values that distinguish between different elements or subsets of elements and their quantitation.

The aforementioned steps can be accomplished within one to three hours. In comparison, current techniques used for materials accountability require several weeks and must be accomplished off-site at specialized laboratories. Therefore, the presently described TMFD systems will provide an extreme improvement in the speed, accuracy, timeliness and cost of actinide detection.

There are nuances when separately applying the above steps for 30-year and 0.5-year cool down fuel types.

In the 30-year fuel having about 30 GWd/MTU burn up and 3 wt. % enrichment the impact of ²⁴²Cm (162-day half-life) is negligible because its alpha activity would be about 100-fold lower. However, due to decay of ²⁴¹Pu (²⁴¹Pu→²⁴¹Am+β−) a significant accumulation of ²⁴¹Am should be accounted for. Even the ²⁴⁴Cm (17.6-year half-life) activity would not be as dominant, and yet, it would be possible to detect its activity within the mix of nuclides since, the computed representative activity ratios of ²⁴¹Am:²⁴⁴Cm is about 6:1. This means that if ²⁴¹Am is detectable within 1 second for example, ²⁴⁴Cm would be detectable within about 6 seconds on average.

In this instance the relative alpha activity ratios of several key actinides from depletion physics are known (calculated via ORIGEN-S for 3 wt. % enrichment and 30 GWd/MTU) to be: ²⁴¹Am to ²³⁸Pu=about 2:1; ²⁴¹Am to ²⁴⁴Cm=about 6:1; ²⁴¹Am to ²³⁹Pu=about 10:1; and, ²³⁸Pu to ²³⁹Pu=about 5:1.

Using the disclosed centrifugal tension metastable fluid detectors, ²⁴¹Am and ²³⁸Pu and ²⁴⁴Cm can be readily monitored, although this may take more time to detect (i.e., compared with that for ²⁴¹Am). For example, even if the relative activity of ²⁴¹Am alone in the sampled mixture is only about 0.1 Bq in the centrifugal tension metastable fluid detector and the associated activities for the other actinides would be: ²⁴⁴Cm (0.017 Bq=0.1/6); ²³⁸Pu (0.05 Bq=0.10/2); and ²³⁹Pu (0.01 Bq); the mixture activity would be the sum equal to about 0.177 Bq. Therefore, scanning from lower tension to higher values, the time to detect and ascertain the various nuclides would be: about 60 s (=1/0.017) for ²⁴⁴Cm alone; followed with about 15 s [=1/(0.017+0.05)] for ²³⁸Pu and ²⁴⁴Cm; about 6 s [=1/(0.017+0.05+0.1)] for ²⁴¹Am together with ²⁴⁴Cm and ²³⁸Pu, and, theoretically, about 5.65 s [=1(0.017+0.05+0.1+0.01] for ²³⁹Pu together with the other three. This process makes it readily possible to estimate for ²³⁹Pu content both directly (i.e., by actual measurement by scanning the P_(neg) space for threshold P_(neg) requirements for detection of specific energy alpha recoils from various actinides as shown in FIG. 7, first starting with ²⁴⁴Cm activity (at lowest Pneg of about −7 bar), then for ²⁴⁴Cm+²³⁸Pu (at a higher P_(neg) of about −8 bar), then for ²⁴⁴Cm+²³⁸Pu+²³¹Am activity (at a subsequently higher P_(neg) of about −8.1 bar) and then for ²⁴⁴Cm+²³⁸Pu+²³¹Am+²³⁹Pu (at a further subsequent higher P_(neg) of about −8.5 bar), and via confirmatory association with the underlying nuclear physics fuel depletion and isotopic decay i.e., via established laws of nuclear physics as encoded within ORIGEN which dictate that, for a given burn up SNF and cool down history, if one knows of the activity of any one of the key constituents, e.g., ²⁴⁴Cm, then, the other constituent isotopes of interest must also be present in certain proportions—unless some type of diversion has taken place. The method for making such assessments is provided in the bullet list below.

Method for monitoring during reprocessing of actinides from spent nuclear fuel after 30-year cool down as described in Algorithm 1.

-   -   Enrichment; Net Average; Burn up; Cooling Period—The process         starts by collecting information from the nuclear power plant         that is the origin of the SNF, so the initial fuel enrichment of         ²³⁵U is known. Also the core-average burn up of the SNF is known         because this is pre-determined ahead of starting the nuclear         reactor based on a prescribed optimized pattern, and the amount         of time the SNF was held in a cooling pool or dry cask ahead of         transmittal to a reprocessing plant is known. From these three         metrics, only the core-averaged burn up would entail a certain         level of uncertainty due to the fact that the rate of fuel         fission in an actual operating reactor varies over about 1 to 2         years of power generation. However, during reprocessing the very         first step involves combining a distributed burn up pattern into         a combination averaged mass via nitric acid dissolution. This         leads to a measure of uncertainty which must then be addressed         via actual measurement down the processing stages; however, up         front, a good first estimate of the various activities can be         made using the well-established ORIGEN code system.     -   Estimate the amount of actinides using ORIGEN-S/run model—per         the aforementioned information, within seconds, the ORIGEN-S         code system can provide a table of actinides and their relative         activity levels.     -   Determine actinide masses, spontaneous fission (SF)/Neutron         Production, (Alpha, n) Production—Based on the ORIGEN model         predictions, neutron production from spontaneous fission and         also from alpha interactions with mixture elements (principally         with O atoms) can be determined. The total neutron production         represents an estimate of the actinide inventory and vice-versa.     -   Calculate Alpha Activity of actinides using masses from ORIGEN         and T½-Similar to the earlier step for neutron production, by         knowing the elemental composition as predicted by the ORIGEN         code model, and knowing the half-lives for alpha decay for the         actinide elements, the activity of each actinide can be         calculated using the formula: Activity in curies per gram         (Ci/g)=(0.693/T_(1/2))×1.6×10¹³/A, where A=atomic mass of the         actinide of interest and T_(1/2) is the half-life.     -   Measure neutron (including for fission rate) production with         CTMFD (Calibrated with ²⁵²Cf and PuBe) for comparison with         prediction—In this step, a TMFD (e.g., a CTMFD) calibrated for         neutron detection efficiency is used to determine the fission         spectrum and random spectrum neutrons. ²⁵²Cf represents a good         fission spectrum source of neutrons, whereas, PuBe or AmBe         isotope based sources produce random (in time) spectra neutron         sources from (alpha, n) nuclear reactions. Both, ²⁵²Cf and PuBe         or AmBe sources are available in the marketplace with         certifications of their strengths. These sources can be used to         calibrate the CTMFD for efficiency of detection of neutrons in         fission and also mixed type (fission and random) source         environments. The calibrated CTMFD can then be positioned a set         distance away from the front-end vat containing the dissolved         contents of SNF and the intensity of neutron emission can be         measured. This measurement can readily be completed within         minutes using a CTMFD with a sensitive volume in the multi-cc         (about 200-500 cc) range, although an ATMFD could also be used.         The intensity of neutron emission forms a measure of how much         activity has built up during burn up of a given level         Importantly, as mentioned earlier, during this front-end stage,         the background activity from beta-gamma decay are about 10         orders of magnitude greater than that for neutron activity, and         can saturate most commonly available neutron detectors in the         marketplace (e.g., ³He, BF₃, etc.). Hence, known detectors fail         to provide information on actinide activity level from their         neutron signals. TMFD technology, on the other hand, remains         completely blind to gamma-beta radiation when using TMFD fluids         such as acetone, isopentane, perfluorooctane, ethanol, methanol,         trimethyl borate and R-113 and the P_(neg) states are kept above         about −20 bar. The measurement provides at least about 90%         intrinsic efficiency at detecting neutrons with detectors that         are sized to permit about 2 mean free path dimensions for         neutrons entering the TMFD sensitive volume (e.g., about 10 cm         diameter, and 5-10 cm height). For assessment of fission         activity, the CTMFD system may be small, in the range of about 1         cc sensitive volume. The dilution of the SNF solution and         detection for fission activity for deriving the actual ²⁴⁴Cm         content by utilizing P_(neg) in the −1 bar range (wherein it         remains blind to neutrons, alphas, gammas and betas) while         remaining more than 95% efficient for detecting fission         fragments has been discussed earlier. The neutron production         rate is 3.45-fold higher than the ²⁴⁴Cm fission rate and this         number can be used to confirm the detection rate for neutrons as         a whole. For 30y cool down fuel, ²⁴²Cm should have decayed down         considerably such that its alpha activity (which, at 180 day         cool down was 5-fold higher than for ²⁴⁴Cm) is 100-fold lower         than that for ²⁴⁴Cm. This can be readily verified by further         dilution of the fission signature mixture by an additional         100,000-fold such that ²⁴⁴Cm can be detected within seconds but         detection of ²⁴²Cm takes 100-fold longer when the P_(neg) is         increased to about −8 bar. The relative contents of the other         isotopes of interest can be determined thereafter.     -   Re-Calibrate ORIGEN until the measured neutron production agrees         with the ORIGEN estimate—As discussed above, an uncertain         upfront metric for ORIGEN code model calculations involved the         averaged SNF burn up during its residence time within the         nuclear power reactor. Using this averaged estimate, a measure         of actinide activity and hence, neutron output can be predicted.         The higher the burn up, the higher the actinide activity buildup         and consequently, the higher the neutron generation rate. The         aforementioned step of actually measuring for the actual neutron         emission intensity offers a strong basis for correcting for the         average SNF burn up such that the predicted and measured neutron         intensity levels agree. Once this is done, the total actinide         inventory comprising the key (Cm, Am, Pu and U) isotopes is         established. The precise amounts can be confirmed during         subsequent stages and can be tracked for possible diversion.     -   Extract samples from dissolved fuel and dilute to about −0.2         decay per second per cc with acetone or other suitable solvents,         diluting according to ORIGEN estimation of activity—Per the         neutron measured activity based correction, the ORIGEN predicted         alpha activity would be known to a good first order of activity.         Only small (microgram) quantities of in-process fluid mixtures         are required at this stage for extraction and subsequent         dilution in the TMFD host fluid material (e.g., acetone) such         that the desired (per ORIGEN predicted) activity level is in the         5 Bq/cc range. The precise level is unimportant and this value         is used for illustrative purposes since, at 5 Bq/cc, detection         for alpha activity in total could be done within 0.2 seconds.         With a more moderate 0.1 Bq/cc activity level, the detection         would take about 10 seconds on average.     -   Confirm the absence of ²⁴²Cm—A key confirming indicator for a         30y cool down fuel is the relative absence of ²⁴²Cm. This is a         due diligence step and can be carried out by placing the fluid         mixture in the CTMFD and noting for any alpha activity at/around         P_(neg) of about −6.5 bar; there the wait time for detection         should conclusively be greater than a prescribed pre-calibrated         amount that includes cosmic background effects in a 1-2 cc CTMFD         system (e.g., about 60 seconds).     -   Measure the amount of ²⁴⁴Cm and verify for the presence of ²⁴²Cm         at 100-fold reduced activity—Once ²⁴²Cm for the sample is         confirmed as being negligible (i.e., detectable), one now         increases the P_(neg) to about −7 bar through −8 bar to note the         relative activity of ²⁴⁴Cm. As an enhanced control, from a         separate sample from the same reprocessing stream but diluted         less (e.g., if the prior dilution was to 0.2 Bq/cc, this         confirmation sample may be diluted to provide about 20 Bq/cc)         and measured for ²⁴²Cm activity by establishing the P_(neg) of         the CTMFD to about −6.5 bar to −7 bar. The laws of nuclear         physics governing fuel burn up and activity buildup require that         the activity of ²⁴²Cm should now be detectable. This provides an         additional and simultaneous control.     -   Determine the concentration of ²⁴⁴Cm—using the data for activity         for ²⁴⁴Cm measured from the previous step which involved the         illustrative 5 Bq/cc sample.     -   Measure combined amounts of ²³⁸Pu, ²⁴¹Am, and ²⁴⁴Cm—Since the         P_(neg) values for ²³⁸Pu and ²⁴¹Am are close enough (i.e., about         −8 bar) the P_(neg) value of the CTMFD can be increased to −8         bar to −8.5 bar to effectively determine the combined activity         of ²⁴⁴Cm, ²³⁸Pu and ²⁴¹Am.     -   Determine the concentration of ²³⁸Pu and ²⁴¹Am—Subtract the         activity of the combined measurements from that for ²⁴⁴Cm alone         to then derive an estimate for the combined amount of ²³⁸Pu and         ²⁴¹Am.     -   Re-Calibrate and refine ORIGEN-S Model for consistency with         experimental findings on Cm, Am, and Pu—the relative activity         levels of these three isotopes can be compared with the activity         levels predicted by ORIGEN to obtain a more certain estimate of         these levels.     -   Determine [²³⁹Pu] from [²⁴¹Am], [²³⁸Pu], [²⁴⁴Cm] as well as         ORIGEN Code ratios.     -   Cross verify this determination with downstream levels of ²³⁹Pu         (in Pu extraction stream) via active measurement and or CTMFD         sampling with monitoring of ²³⁸Pu and ²³⁹Pu

Ratio [in activity/MTU; 30 y cool down; 30 GWd/MTU; Actinides with 3 wt. % enrichment ²⁴¹Am/²³⁸Pu 2:1 ²⁴¹Am/²⁴⁴Cm 6:1 ²⁴¹Am/²³⁹Pu 10:1  ²³⁸Pu/²³⁹Pu 5:1

Method for monitoring of actinides during reprocessing for spent nuclear fuel after a 180-day cool down. The following steps are essentially the same as that discussed above for 30-y cool down fuel. The principal exceptions being that, for the 180 day cool down fuel, the alpha rate is dominated by ²⁴²Cm, the fission neutron rate is dominated by ²⁴⁴Cm (with this rate being about 5-fold greater than that from ²⁴²Cm), and the alpha and fission neutron emission rates from ²⁴¹Am are negligibly small.

-   -   Enrichment; net average burn up; cooling period.     -   ORIGEN-S/Run Model on PC.     -   Actinide Masses SF Neutron Production (Alpha, n) Production.     -   Calculate alpha activity of actinides using masses from ORIGEN         and T½.     -   Measure neutron production with CTMFD (Calibrated with ²⁵²Cf and         PuBe) for comparison with prediction.     -   Re-Calibrate ORIGEN until it agrees with measured values.     -   Extract samples from dissolved fuel and dilute to about −0.2         decays per second (for fission activity monitoring and         separately, also for alpha activity monitoring) with acetone         according to ORIGEN estimation of alpha-activity. Acetone is a         convenient universal solvent liquid to choose; however, other         suitable liquids include (e.g., ethanol, methanol, R-113,         isopentane, etc.) and can be used so long as the SNF mixture is         soluble.     -   Measure for ²⁴²Cm activity from alpha activity monitoring; for         ²⁴⁴Cm from fission activity.     -   Determine concentration of ²⁴²Cm from measured activity.     -   Measure combined ²⁴²Cm and ²⁴⁴Cm.     -   Determine concentration of ²⁴⁴Cm.     -   Measure combined ²³⁸Pu, ²⁴²Cm, and ²⁴⁴Cm.     -   Determine concentration of ²³⁸Pu.     -   Measure combined Am-241, ²³⁸Pu, ²⁴²Cm, and ²⁴⁴Cm.     -   Determine concentration of ²⁴¹Am.     -   Re-Calibrate and refine ORIGEN-S model for consistency with         experimental findings on Cm, Am, and Pu.     -   Determine concentrations of ²³⁹Pu from ²⁴¹Am, ²³⁸Pu, ²⁴²Cm, and         ²⁴⁴Cm as well as ORIGEN Code ratios.     -   Cross verify with downstream levels of ²³⁹Pu in the Pu         extraction stream via active monitoring and/or CTMFD sipping         based monitoring of ²³⁸Pu and ²³⁹Pu.

Ratio of alpha activities (with 5 wt. % enrichment, 50 GWd/MTU Actinides burn up, 0.5 y cool down) ²⁴⁴Cm/²³⁸Pu 1:1 ²⁴²Cm/²⁴⁴Cm 5:1 ²⁴²Cm/(²⁴⁴Cm + ²³⁸Pu) 2:1 ²³⁸Pu/²³⁹Pu 10:1 

Compared to 30-year cool down Spent Nuclear Fuel, the ²⁴¹Am content in 0.5 year cool down Spent Nuclear Fuel is negligible, but the impact of ²⁴²Cm should be included because its 0.5y half-life would not have allowed significant decay. The relative alpha activity ratios from depletion are as follows: ²⁴⁴Cm to ²³⁸Pu=about 1:1; ²⁴²Cm to ²⁴⁴Cm=about 5:1; ²⁴²Cm/(²⁴⁴Cm+²³⁸Pu)=about 2.5:1; and, ²³⁸Pu to ²³⁹Pu=about 10:1. The neutron activity levels (n/s/MTU) from spontaneous fission are dominated by ²⁴⁴Cm as noted: ²⁴¹Am (9.8×10); ²⁴²Cm (about 1.3×10⁸); ²⁴⁴Cm (about 5×10⁸); ²³⁸Pu (about 5.5×10⁵); ²³⁹Pu (89) and ²⁴⁰Pu (2.5×10⁶).

Furthermore, as noted above, the total neutron emission rate of about 5×10⁸ n/s/MTU is largely from Cm with the intensity ratio based on Spontaneous Fission half-lives ²⁴⁴Cm to ²⁴²Cm about 5:1. Interestingly, ²⁴²Cm activity while not as high as ²⁴⁴Cm is readily discernible from the activity levels of ²⁴⁴Cm and also ²³⁸Pu. Since the activity of ²⁴¹Am is negligible, there is less chance for interference when monitoring for ²³⁸Pu with its closely spaced alpha energy emission; therefore, the quantity of ²³⁹Pu is more confidently obtained for 0.5-year cool down Spent Nuclear Fuel compared with 30-year cool down Spent Nuclear Fuel. The methods for such assessments are provided in the following paragraphs.

The sensor system and structure comprising Tension Metastable Fluid Detector sensor hardware are combined with ORIGEN-S based simulation and prediction methods for monitoring Pu, U, and other actinide isotopes, at initial processing stages (as described above) and through subsequent stages in a chemical nuclear reprocessing plant. For tasks that involve mixing the SNF bearing solution with the TMFD fluids, such as for direct monitoring of fission and alpha rates, dilution will be necessary. The degree of dilution must be estimated. For this, the following series of steps can be used:

-   -   The gamma-beta activity (A) in SNF can be estimated from the         formula, A (Bq)=P×10⁶×[t^(−0.2)−(t+T)^(−0.2)]×3.7×10¹⁰, where, P         is the thermal power in megawatts, T (in days) is the duration         of operation of the reactor at that power level, and “t” is the         time (in days) after shutdown of the reactor. For example, for a         typical 3,000 MW reactor, shutdown after operating for 18         months, at “t=180” days after shutdown will have gamma-beta         radioactivity A of about 10¹⁹ Bq. This is the gamma-beta         activity in the SNF which is typically about 40 tons, although         this can vary by about 2-fold. Therefore, the beta-gamma         activity per MTU would be about 2×10¹⁷ Bq/MTU. Since this         activity is mainly gamma-beta variety, and gamma photons can be         readily measured by conventional detectors such as NaI, the         level of activity per MTU on average can be estimated even at         the front end—note: this does not provide any reasonable         information relating to the actinide content. Since the density         of UO₂ is about 10⁵ kg/m³, the volume per MTU of SNF prior to         dissolution would be about 40×10 3/1×10⁵ or, about 0.4 m³.         Assuming the SNF is diluted by 100-fold, the specific activity         (gamma-beta) per MTU would amount to about 6×10¹¹ Bq/MTU/cc     -   Once the total activity level per unit volume in the dissolved         SNF vat is known the dissolved SNF must be diluted for         gamma-beta blind neutron, fission and alpha activity monitoring.         For 180 cool down fuel: the alpha activity per MTU is estimated         to be about 10¹⁵ Bq/MTU; the neutron activity level dominated by         Cm isotopes is about 5×10⁸ neutrons/sec/MTU (from which the         fission activity is readily obtained by dividing by the         multiplicity factor for Cm which is about 4, to result in         fission activity rates of about 10⁸ fissions/sec/MTU). The alpha         activity is about 200 times lower than the gamma-beta activity,         whereas, the fission activity is about 2×10⁹ times lower than         the gamma-beta activity in the SNF vat.     -   For efficient detection in the TMFDs the dilution can be         performed with acetone. The TMFD detection capability with a         given detection fluid (e.g., acetone) can be significantly         reduced if significant quantities of inorganic fluids such as         water or nitric acid are used. The addition of about 1 to 5 vol.         % of inorganic liquids does not impair performance if the TMFD         fluid is acetone. This additional volume of inorganic substances         may be increased even to about 10 vol. % by mixing isopentane         with acetone. However, for SNF vat mixtures of the type         discussed above the inclusion of SNF bearing inorganic fluids         such as HNO₃ should not pose an issue. For TMFD systems meant to         monitor for fission rates and alpha rates, the TMFD's sensitive         volume may be kept low (e.g., 1 cc) since the track lengths of         alphas and fission fragments can be expected to be no more than         a few tens of microns at most. Also, for detection within a         reasonable amount of time (e.g., within 10 seconds on average),         the fission or alpha activity per cc should then be about 0.1         Bq/cc. This implies that, for 180-day cool down fuel, under the         above-mentioned conditions, and assuming that we add 1 vol. % to         the TMFD volume, the SNF in the vat must be diluted by a factor         of: for alpha activity monitoring dilute by about 200/0.01 or         2×10⁴; for fission activity monitoring dilute by about         2×10⁹/0.01 or about 2×10¹¹.

An optimal reprocessing system for generating new fissile fuel for energy generation should efficiently and securely separate key elements such as U, Cm, Pu into various radioactive waste streams. For example, from FIG. 1 one (UREX) stage involves removal of U and ⁹⁹Tc from the balance of fission products and transuranics. It is important to ensure that such separation occurs as intended and that quantities of actinides such as Cm and Pu are not inadvertently diverted into this stream. Therefore, in U/Tc bearing streams, monitoring for ²³⁵U, ²³⁸U and ⁹⁹Tc helps to ensure the “absence” of Cm, Pu, Am-type actinides.

An assay for the various isotopic separations is described. The method utilizes instrumentation for conducting assays in real time. In one method a Tension Metastable Fluid Detector monitoring system can include at least two detector banks. The first bank can include a calibrated Tension Metastable Fluid Detector for monitoring neutrons from spent fuel spontaneous fission and α-n reactions, where the predicted intensity is calculated to be in the range of about 2.5×10⁸ n/s/MTU, and about 5×10⁸ n/s/MTU for 30-year (˜30 GWd/MTU burn up and 3 wt. % enrichment) and 0.5-year (about 40 GWd/MTU burn up and 5 wt. % enrichment) cool down Spent Nuclear Fuels, respectively. Both fuel sources are dominated by ²⁴⁴Cm. The relative contribution to neutron production from α-n reactions varies from about 3% for 30y cool down fuel to about 10% for 0.5y cool down fuel. Monitoring can provide a basis for estimating the quantity of ²⁴⁴Cm and the amount of the rest of actinides of interest can be calculated from this using the disclosed methods. The neutron detector bank would preferably include 2 CTMFDs with sensitive volumes in the 300-500 cc range and using detection fluids such as isopentane and trimethyl borate so that the instrumentation can monitor with at least 90% intrinsic efficiency. Although one such CTMFD would suffice, having two CTMFDs should enable backup detection and/or to cross-check the reading of the other.

In an embodiment the second bank can contain at least 4 Tension Metastable Fluid Detectors each operating at tension metastable states connected with detection of key isotopes ²⁴⁴Cm, ²⁴²Cm, ²³⁸Pu, and ²⁴¹Am. Considering that this second bank pertains to detection of fission fragments and also for alpha emitters via sampling and dilution, the central sensitive volume of the CTMFD may be as small as 1 cc. This is because the fission fragments and alpha particles and recoil atoms get readily stopped within the detection fluid within several tens of microns. A sampling system can be used to draw a quantity of fluid (in μL to mL volumes) from the mixture vat and to dilute the sample to the 0.1 Bq range, prior to introducing the mixture into the Tension Metastable Fluid Detector systems for assessment and detection of the presence of specific key isotopes (²⁴²Cm, ²⁴⁴Cm, ²³⁸Pu, ²⁴¹Am and ²³⁹Pu). As explained earlier, the degree of dilution for alpha activity in the 10¹⁴ Bq/MTU and for fission activity in the 10⁸ Bq/MTU will necessarily be different. The second bank of detectors provides for a relatively exact (less than 1% error margin) estimate for the relative quantities of ²³⁹Pu and with greater error (approximately 10%) for ²³⁵U in the mixture for reasons explained earlier. These two types of detector banks can form the basic setup for various branches of the reprocessing stream. This is because the alpha-neutron producing dominant Cm isotopes are the last to be removed. At each intermediate stream, actinides are selectively isolated and additional detector banks or even active interrogation may be utilized (e.g., for monitoring for ²³⁵U which is a relatively weak alpha and neutron emitter).

The U/Tc extraction (UREX) line normally contains negligible quantities of transuranic isotopes. The U isotopes would primarily be ²³⁴U, ²³⁵U and ²³⁸U, and since ⁹⁹Tc is a beta emitter with about a 0.2 million year half-life which TMFDs do not detect, the measurement is for only the three U isotopes. Due to extremely high Spontaneous Fission half-lives of over 10¹⁵y and consequently a very low level of fission activity, and also for alpha-emission half-lives of over 10⁷y in relation to the amount of uranium neutron production in this UREX extraction stream can be considered to be similar in intensity as background radiation. Such a UREX extraction line cannot be readily monitored for neutron activity via passive neutron detection unless unintentional diversion takes place for other actinide elements such as Pu. For such instances, the presence of significant neutron activity is a tell-tale sign and the two aforementioned TMFD bank types passively monitoring the UREX line would detect such an event. Active external neutron induced fission based monitoring does indeed offer such a possibility for active interrogation to decipher for ²³⁵U content. We have ascertained that either a 1 Ci Pu—Be isotope source or using an equivalent output of about 10⁶ n/s, 14 MeV D-T pulsed generator source neutrons can measure about 100 g quantities of U from fast neutron-induced fissions for determining the quantity of ²³⁸U and ²³⁵U within minutes of monitoring. The quantity of ²³⁵U in Spent Nuclear Fuels in this process stream can thus be determined in the mix but would require an external neutron source. In order to determine the quantity of ²³⁵U in Spent Nuclear Fuels a 1 Ci Pu—Be, equivalent ²⁵²Cf, or accelerator-driven sources can be used together with a down-scattering medium such as paraffin or polyethylene of about 10 mean free path lengths (e.g., 0.2 m) thickness. Such down-scattering is useful because the probability of fission (the cross-section) increases logarithmically with reduced neutron energy, rising to about 600 barns for ²³⁵U fission, versus only about 1 barn at 14 MeV levels. This can then be cross-checked using the sampling technique after dilution to provide direct information on alpha activities of ²³⁴U (4.77 MeV), ²³⁵U (4.58 MeV) and ²³⁸U (4.2 MeV) based upon the data shown in FIG. 4. The two independent checks of neutron and alpha radiation provide accurate monitoring of ²³⁵U content within 30 minutes to an hour. Sweeping the P_(neg) tension pressures from about −9 bar to −10 bar (about 10% change) requires only the increase in rotational speed of the CTMFD system by about 3% since P_(neg) levels vary quadratically with rotational speed and neutron output monitoring in the absence of active monitoring. This can be used to determine if material diversion is taking place since diversion of Cm/Pu/Am content even at about 0.1% would give rise to unmistakable signatures of neutron-alpha activity in the U/Tc stream because the P_(neg) values required for detection of Cm/Pu/Am isotopes occurs at lower P_(neg) values than that required for detecting U isotopes (FIGS. 4 and 7).

For a Cs—Sr extraction stream (FPEX) as shown in FIG. 1 with a mixture that presumably includes only beta-gamma emitters and no detectable neutron or alpha emitters, a single Tension Metastable Fluid Detector sensor having a sensitive volume in the range of about 300-500 cc, with an isopentane detector fluid, set to operate at P_(neg) of around −5 bar can be used to reliably monitor for neutron activity above cosmic background levels. This can be used to provide an indication of the presence or absence of transuranic actinide (e.g., Pu) diversion. Alternatively, a two detector bank of TMFDs may also be utilized since the added cost and complexity with use of TMFDs is very low in comparison with known methods which require off-site shipment of liquid streams.

Background beta-gamma activity levels in the subsequent NPEX stream composed of Pu and Np elements, are significantly lower, for example about 1000-fold lower or less, and since Tension Metastable Fluid Detectors are already blind to beta-gamma radiation, the monitoring system would be the same as that described previously for the front end monitoring. A Tension Metastable Fluid Detector preferably having about 300-500 cc of detector fluid which is preferably isopentane at a P_(neg) of about −6 bar could be used to monitor for neutrons. This can be coupled with a bank of 3 or more Tension Metastable Fluid Detectors (about 1 cc; acetone filled, P_(neg) varying from −7 bar to about −9 bar) to simultaneously monitor for ²³⁸Pu, ²³⁹Pu, and ²³⁷Np). A similar embodiment could be used to monitor a TRUEX extraction stream comprising a balance of the transuranic elements or any similar set of TMFDs for neutron and alpha monitoring for ²⁴²Cm, ²⁴⁴Cm, and ²⁴¹Am isotopes. This comprehensive monitoring system provides the advantage of allowing cross-checks in real-time with the measurements upfront to detect a diversion of Special Nuclear Materials.

While the above description is provided for monitoring of fissile isotopes such as ²³⁵U and ²³⁹Pu in extremely high gamma-beta fields for the UREX reprocessing scheme, the approach is general in nature and may be readily employed for selective neutron-alpha-fission activity monitoring, for other reprocessing schemes such as PuREX and for monitoring for storage and flow of SNMs within weapons manufacturing, deployment and stockpile facilities. In addition the detector systems have the ability to provide information about the direction from which the neutrons originate and also for imaging the sources and multiplicity as shown in FIG. 6 to identify what specific fissile isotope is present since fission of ²³⁵U releases about 2.5 neutrons/fission versus ²⁵²Cf which releases about 4 neutrons/fission. Therefore, the ATMFD technology may be deployed in numerous situations where neutron detection is needed and can be used to track the location and movement of fissile nuclei.

Table 4 summarizes the types of radiation that can be detected by the disclosed compositions and methods.

TABLE 4 Key Radiation Signatures Detectable by Tension Metastable Fluid Detector Systems Signature Discussion Alpha Energy U and Transuranic isotopes are intimately mixed in process solutions. The Spectra gamma blind Tension Metastable Fluid Detector is capable of discerning the (dissolved presence of isotopes ranging from ²³⁸U to ²³⁹Pu to ²⁴²Cm. solutions) For deciphering isotopes in mixtures, convert the time vs. P_(neg) information in FIG. 4 into quantitative measures of the actinide concentration in solutions using the radioactive rate constant additive principle. Neutron Energy Tension Metastable Fluid Detectors are capable of direct detection of thermal Spectra and fast neutrons (FIG. 6). This direct measurement is a remarkable (~0.01 eV to 10 improvement over current methods that use multiple thermal neutron detectors MeV energy) to infer spectral information. For reprocessing plants, the disclosed TMFDs could be used to scan (in minutes) the time vs. P_(neg) information and thereby quantify an isotopic specific neutron spectrum. Neutron A 10 cm diameter TMFD is able to detect and track an 8 kg ²³⁹Pu source Directionality within 30 s to within 10° with 80% confidence up to 25 m away. The disclosed (source location TMFDs can be used to monitor, in virtual real-time, the flow of neutron- discernment) emitting Special Nuclear Materials within various piping streams of chemical processing plants. Neutron Fissile isotopes such as 235U and 239Pu may be detected using the active Multiplicity interrogation methods disclosed herein. In addition, neutron multiplicity data (Direct can be characterized. Existing systems rely on banks of multiple 3He Observation of detectors. ATMFD systems can decipher multiple coincident events within a Fission) single detector or a bank of detectors to decipher the isotope in question based on multiplicity. Fission Fission fragments typically originate with energies ranging from 80 to 100 Product Recoil MeV on average. A TMFD using acetone as a detection fluid requires −7 to −9 (Dissolved Fissile bar to detect 1 to 5 MeV neutrons or alpha particles and only −0.2 bar to detect Actinides) 80 to 100 MeV FFs, as shown in FIG. 7. This constitutes a signature for the presence of fissile materials.

The neutron detection methods and system for neutron detection can be better understood with reference to FIGS. 7 and 13. FIG. 13 provides an illustration of a detection system for detecting neutrons from a sample 100. having five TMFD detectors 10, 20, 30, 40 and 50 that are set to distinct P_(neg) values such that neutrons originating from different elements teat have different energies can be distinguished. For example, if detection of neutrons from each of ²³⁸U, ²³⁴U, ²³⁹Pu, ²⁴¹Am, ²³⁸Pu, ²⁴⁴Cm, and fission events are desired, detector 10, for example, could be set with a Pneg of −10 bar according to FIG. 7. To detect ²³⁴U, ²³⁹Pu, ²⁴¹Am, ²³⁸Pu, ²⁴⁴Cm, and fission events, detector 20 could be set to have a P_(neg) of greater than −9.5 bar and less than −9 bar. To detect ²³⁹Pu, ²⁴¹Am, ²³⁸Pu, ²⁴⁴Cm, and fission events, detector 30 could be set to have a P_(neg) of greater than −9.0 bar and less than −8.5 bar. To detect ²⁴¹Am, ²³⁸Pu, ²⁴⁴Cm, and fission events, detector 40 could be set to have a P_(neg) of greater than −8.5 bar and less than −8.1 bar. To detect ²³⁴U, ²³⁹Pu, ²⁴¹Am, ²³⁸Pu, ²⁴⁴Cm, and fission events, detector 20 could be set to have a P_(neg) of greater than −9.5 bar and less than −9 bar. To detect ²⁴⁴Cm, and fission events, detector 50 could be set to have a P_(neg) of greater than −8.0 bar and less than −7 bar. A detector could also be set to have a Pneg of between about −0.2 and −7 bar for detection of ²³⁵U fission. The detectors can communicate with a computer 200 that can be used to calculate the amounts of various neutron sources present in sample 100. The computer can also be implemented with ORIGEN code and used to carry out the calculations of actinide species present in spent nuclear fuel. 

The invention claimed is:
 1. A method for monitoring actinides during reprocessing of spent nuclear fuel after 30-year cool down comprising: enriching a spent nuclear fuel sample derived from burned nuclear fuel having a specific initial enrichment, estimating the amount of actinides in the sample, determining actinide masses and spent fuel neutron production from fission and a, n production, predicting fission and a activity of actinides using estimated masses and T½, measuring neutron production with a TMFD, comparing measured neutron and fission production with predicted neutron and fission production repeating the predicting step until the numbers agree within about 10% or less of the measured neutron production, removing a sample from the dissolved fuel, diluting the sample to about 0.1 to about 10 decay per second to allow detection in the TMFD system within about 5 to about 60 seconds, estimating the radiation activity of the sample, confirming absence of ²⁴²Cm activity, measuring ²⁴⁴Cm and determining the concentration of ²⁴⁴Cm in the sample measuring the combined ²³⁸Pu, ²⁴¹Am, and ²⁴⁴Cm and determining the concentration of ²⁸³Pu, ²⁴¹Am in the sample, re-calibrating the actinide concentration estimates to correspond with measured amounts of Cm, Am, and Pu within about 10% or less, determining the concentrations of ²³⁹Pu from ²⁴¹Am, ²³⁸Pu, ²⁴⁴Cm and estimated ratios, cross verifying the amounts of ²³⁹Pu from extraction stream using active TMFD measurement.
 2. A method for monitoring actinides during reprocessing of spent nuclear fuel after 180-day cool down comprising: enriching a spent nuclear fuel sample, estimating the amount actinides in the sample, determining actinide masses and spent fuel neutron production from fission and α, n production, predicting fission and α activity of actinides using estimated masses and T½, measuring neutron production with a calibrated TMFD, comparing measured neutron and fission production with predicted neutron and fission production, if the measured and predicted neutron production numbers do not agree repeat the predicting step until the numbers agree within 10%, removing a sample from the dissolved fuel, diluting the sample to about 0.1 to about 10 decay per second to allow detection in the TMFD system within about 5 to about 60 seconds, estimating the radiation activity of the sample, measuring ²⁴²Cm and determining the concentration of ²⁴²Cm in the sample, measuring combined ²⁴⁴Cm and ²⁴²Cm and determining the concentration of ²⁴⁴Cm in the sample, measuring the combined ²³⁸Pu, ²⁴²Cm, and ²⁴⁴Cm and determining the concentration of ²³⁸Pu, in the sample, measuring combined 238Pu, ²⁴¹Am, ²⁴⁴Cm and ²⁴²Cm and determining the concentration of ²⁴¹Am in the sample, re-calibrating the actinide concentration estimates for consistency with measured amounts of Cm, Am, and Pu, determining ²³⁹Pu from ²³⁸Pu, ²⁴¹Am, ²⁴⁴Cm and ²⁴²Cm measurement and estimated ratios, cross-verifying the amounts of ²³⁹Pu by either active TMFD monitoring or TMFD sipping based monitoring of ²³⁸Pu and ²³⁹Pu. 